The role of FeCrAl alloys in ensuring radiation resistance and safety of nuclear reactors

Abstract

In order to develop accident-tolerant fuel (ATF) cladding for nuclear reactors, FeCrAl alloys are considered a key material in the nuclear energy sector. These alloys demonstrate excellent resistance to corrosion and oxidation under high temperatures and aggressive environments. Moreover, they exhibit thermal conductivity comparable to the zirconium-based alloys currently in use. However, during nuclear reactor operation, exposure to ionizing radiation can alter the mechanical properties of these materials, leading to irradiation-induced hardening. This article provides an overview of the primary experimental techniques—such as nanoindentation and microhardness testing - and theoretical approaches including the Nix-Gao model, dispersion barrier hardening (DBH) theory, density functional theory (DFT), molecular dynamics (MD), and discrete dislocation dynamics (DDD), used to study the hardening process in FeCrAl alloys after irradiation. It also discusses the effects of microstructural changes, such as the formation of dislocation loops and α'-precipitates. Strategies to enhance radiation resistance through the incorporation of oxide dispersion-strengthened (ODS) particles are also described. Theoretical models help bridge atomic-scale defects with macroscopic mechanical behavior. Future research directions include alloy composition optimization, long-term performance assessment, and the application of additive manufacturing technologies.